In South Korea, the replacement of steam generators began with Kori Unit 1 in 1995, and to date, 20 steam generators have been replaced and are currently stored in intermediate storage facilities. In the future, additional decommissioned steam generators may arise due to measures like the extension of the lifespan of nuclear power plants. In Korea, technological development for dismantling steam generators is underway, and there is no track record of actual dismantling. Although the replaced decommissioned steam generators are stored in intermediate facilities, for site recycling purposes, steam generators, which have relatively lower radiation doses compared to reactor heads and other primary equipment, should be prioritized for dismantling. While there are various specifications for steam generators, those dismantled and stored domestically are of the Recirculation Type. They can be classified into three types: the Westinghouse type WH-51 used in Kori Unit 1, the Fra-51B used in Han-ul Units 1 and 2, and the OPR-1000 used in Han-ul Units 3 and 4. The quantity of U-Tubes varies depending on the specification, but the radiation is concentrated in the primary side components, the U-Tube and Chamber. Since the parts related to the secondary side are not contaminated, they can be disposed of independently after classification. To dismantle a steam generator, it is of utmost importance to first create a scenario regarding where and how the dismantling will take place. Through the analysis of the advantages and disadvantages of each scenario, the optimal timing, location, and cutting method for dismantling should be researched. Furthermore, based on those findings, the best scenario should be derived through an analysis of worker radiation exposure and dismantling costs. To achieve this, a 3D simulation software developed by Cyclelife Digital Solutions under the French EDF was utilized to conduct simulations based on different dismantling schedules and methods. As a result, the optimal scenario for dismantling the steam generator was derived.

Among the twenty six nuclear power plants in Korea, twenty four plants are currently in operation excluding the two permanently shut-down Kori #1 and Wolsung #1 plants. The decommissioning process includes many tasks such as cutting, decontamination, disposal and treatment. Among the tasks, because cutting is one of the tasks performed close to the target structure, there is a possibility for the workers to be exposed excessively to the radiation. There are representative large structures such as steam generators, nuclear reactors, reactor coolant pump, and pressurizer, made of metals, and radioactive concrete, made of concrete. Especially, compared to the trend of research to manage the radiation exposure of steam generators that are directly connected to pressurizers, the trend of research to manage the radiation exposure of pressurizers to workers is not satisfactory. Moreover, although there have been many studies on radioactive concrete, the studies to manage the radiation exposure to workers with a systematic cutting scenario are insufficient. In this study, radioactive concrete, a representative large structure made of concrete, was selected as the target for evaluation. The conditions for evaluation were cutting speed (1~10 m2/hr) and the time for cutting (permanent shutdown~30 years after the shutdown). A cutting scenario was developed by applying the situation for abrasive decontamination beforehand and Hot-to-Cold and Cold-to-Hot, and effort was made to derive a reasonable plan. The evaluation result derived were hourly radiation dose distribution of 1.19~0.103 mSv/hour and 1.29~0.0113 mSv/hour for a scenario without abrasive decontamination (in the order of Hot to Cold, Cold to Hot), and hourly radiation dose distribution of 0.547~0.0479 mSv/hour and 0.608~0.0522 mSv/hour for a scenario with abrasive decontamination. The maximum value of collective dose derived was 1.54E+04 mSv at the cutting time of permanent shutdown with cutting speed of 1 m2/hour in the Cold to Hot scenario before abrasive decontamination, and the minimum value derived was 5.15E+01 mSv at the cutting time of 30-year after the permanent shutdown with cutting speed of 10 m2/hour in the Hot to Cold scenario after abrasive decontamination.

Decommissioning of a nuclear power plant (NPP) generate large amounts of various types of wastes. In accordance with the Nuclear Safety and Security Commission Notice of Korea (No. 2020- 6), they are classified as High Level Waste (HLW), Intermediate Level Waste (ILW), Low Level Waste (LLW), Very Low Level Waste (VLLW) and Exempt Waste (EW) according to specific activities. More than 90% of the wastes are at exempt level, mostly metal and concrete wastes with low radioactivity, of which the concentrations of nuclides is less than the allowable concentration of self-disposal. The self-disposal or recycling of these wastes is widely used worldwide. More than 10,000 drums, based on 200 L drum, are expected to be produced in the decommissioning process of a unit of nuclear power plant. Due to the limited storage capacity of the intermediate & low level waste disposal facility in Gyeongju, recycling and self-disposal of EW are actively recommended in Korea. A variety of scenarios were proposed for recycling and self-disposal of decommissioning metal/ concrete wastes, and a computational program called REDISA was developed to perform the dose evaluation for each recycling and self-disposal scenario. The REDISA computer program can calculate external and internal exposure doses by simulating the exposure pathways from waste generation, thru transport, processing, manufacture, to the final destination of recycling or self-disposal. In this study, the self-disposal scenario was only considered for the dose evaluation. Many studies have been conducted to evaluate the exposure doses of the radioactive waste disposal sites. However, there have been few researches on dose evaluation for self-disposal landfills. In particular, the dose evaluation is important not only during the operation period, but also for a long period after the facility is closed. To this end, we developed a conceptual model for dose evaluation for post-closure scenarios of the self-disposal landfill of decommissioning metal/concrete wastes with reference to the methodology of IAEA-TECDOC-1380. The model incorporates three exposure pathways, including external exposure from contaminated soil, internal exposure by inhalation, and internal exposure by ingestion of water and food grown in contaminated soil. The duration of the dose evaluation is set to 100,000 years after the closure of landfill facility. Co-60 was selected as dominant nuclide, and dose evaluation was performed based on unit specific activity of 1 Bq/g. Exposure doses shall be verified for their application in accordance with the annual dose limit of 10 Sv/yr for self-disposal. As a result, the post-closure scenario of selfdisposal landfills have shown negligible effects on public health, which means that the exposures doses from transportation and operational processes should be considered more carefully for selfdisposal of decommissioning metal/concrete wastes.